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The integral thermal-hydraulic test facility is an important part of the design certification and safety review of nuclear power plants. During the past few decades, a considerable number of resources and efforts have been devoted to establishing integral test facilities and carrying out experimental programs throughout the world. These test facilities provide a large database for computer code assessments and a better understanding of thermal-hydraulic phenomena for postulated accident transients. In this paper, some main integral thermal-hydraulic test facilities are summarized from the perspective of scaling design and applicability evaluation. Scaling analysis methods are presented for the rational design of scaled-down test facilities. The main characteristics of these thermal-hydraulic test facilities are illustrated, including their construction history, scaling design, test matrix, verified codes and related research works. The scaling distortion and experimental applicability of different integral test facilities are compared and evaluated. This review is expected to provide the reference guidance for the rational design and application evaluation of integral thermal-hydraulic test facilities.

The RELAP5/MOD3.3 code was used to simulate a Main Steam Line Break (MSLB) Loss of Coolant Accident (LOCA) on a Generation III type Reactor with passive safety features. The MSLB is a hypothetical multi-failure large break LOCA accident scenario. The results of the base case calculation have been compared with the available integral testing data to assess the code’s capability to predict the large break LOCA phenomena in this generic reactor, particularly with passive safety systems. Various flows such as break flow and gravity injection flow were simulated with a high level of accuracy. In addition, both sensitivity cases and uncertainty analysis were carried out to build the best estimate plus uncertainty practice. The sensitivity study was performed with different initial conditions and adjusted containment cooling capacities. The results confirmed that the code model held its stability and repeatability against a few variations of initial and boundary conditions. In the uncertainty quantification process, the 95/95 uncertainty bands of key output parameters were obtained using the Wilks statistic methods. Ten input parameters were selected with imposed uncertainty distribution to perform uncertainty propagation analysis. The contribution weight of input parameters for the uncertainty of minimum core level were identified with the Spearman rank correlation coefficient in sensitivity analysis.

CAP1400 has been developed as a new passive nuclear power plant in China based on AP1000 technique. In order to support the CAP1400 licensing, a scale-down integral test facility named ACME (Advanced Core-cooling Mechanism Experiment) has been designed and constructed. In this study, best-estimate calculation using RELAP5/MOD3.4 code plus uncertainty quantification and sensitivity analysis of the small break loss of coolant accident (SBLOCA) transient were carried out for the ACME test facility. The important thermal-hydraulic phenomena during the SBLOCA transient were investigated through the comparing calculations between the ACME and CAP1400. Thereafter, uncertainty quantification process was demonstrated including the selection of input uncertain parameters, the application of Wilks’ nonparametric statistics and the propagation of uncertainty through the SNAP interface coupling RELAP5 and DAKOTA codes. Furthermore, the results of uncertainty and sensitivity analysis were discussed. Through the comparison of the thermal-hydraulic modeling results between ACME and CAP1400, the results show that the ACME test facility is able to reproduce the important thermal-hydraulic phenomena of CAP1400 by reasonable scaling design. The 95/95 uncertainty bands of some key output parameters are obtained through uncertainty quantification. The lower band of the minimum core level is maintained still above the top of the active core section, thus the core uncover will not occur with 95% confidence level. Important input uncertain contributors for the minimum core level are identified through the measure of Spearman rank correlation coefficients in sensitivity analysis. The demonstration of best-estimate plus uncertainty analysis (BEPU) application will provide some guidance for the safety analysis of advanced passive nuclear power plants.

In the scaled-down test facilities that simulate the accident transient process of the prototype nuclear power plant, the stored energy release in the metal structures has an important influence on the accuracy and effectiveness of the experimental data. Three methods of stored energy analysis are developed, and the mechanism behind stored energy distortion in the test facilities is revealed. Moreover, the application of stored energy analysis is demonstrated for the ACME test facility newly built in China. The results show that the similarity requirements of three methods analyzing the stored energy release decrease gradually. The physical mechanism of stored energy release process can be characterized by the dimensionless numbers including Stanton number, Fourier number and Biot number. Under the premise of satisfying the overall similarity of natural circulation, the stored energy release process in the scale-down test facilities cannot maintain exact similarity. The results of the application of stored energy analysis illustrate that both the transient release process and integral total stored energy of the reactor pressure vessel wall of CAP1400 power plant can be well reproduced in the ACME test facility.

In order to investigate the performance of a Generation III BWR safety system, an Integral test facility was designed and conducted, and several integral tests were performed. The best estimate safety analysis of Main Steam Line Break (MSLB) accident using RELAP5/MOD3.3 and RELAP/SCDAPSIM/MOD3.4 (from now on referred as RELAP/SCDAP3.4) was performed. By comparing the simulation results with the test results, the prediction capability of the RELAP5/MOD3.3 and the RELAP/SCDAP3.4 was compared and analyzed. The RELAP/SCDAP3.4 code was selected to perform the uncertainty study. According to the Phenomenon Identification and Ranking Table, 12 input treatable parameters and 29 source code coefficients were selected with imposed uncertainty distribution to conduct uncertainty analysis. The 95/95 uncertainty band of the minimum core collapsed level were obtained and analyzed. In addition, the contribution weight of input treatable parameters and source code coefficients for the minimum core collapsed level were identified with the Spearman rank correlation coefficient.

Solar steam generation technology can utilize abundant and renewable solar energy for many applications. In this work, we proposed a low-cost high-efficiency solar steam generator by using wick materials with wrapped graphene nanoparticles, and the energy efficiency can reaches up to 80%. Instead of traditional smearing method, the chemical wrapping method was used for better coating the graphene nanoparticles on the wick materials. Through the SEM morphological results, the graphene nanoparticles are shown to be evenly wrapped across the fibres of the wick material, which have better dispersity and stability. The evaporation rate, instantaneous energy efficiency and the solar absorptivity of three wick materials with/without nanoparticles under different conditions were compared and analyzed. Among the three different wick materials, the fine fluffy cloth can provide more three-dimensional contact area for wrapping graphene nanoparticles and thus contribute to better evaporation. Additionally, the influence of two different reduction methods and different concentrations of graphene oxide solution on the energy efficiency was also analyzed. Our work offers a simple and effective way of using nanotechnology in practical application for better solar-to-heat conversion.